JPH0320718B2 - - Google Patents

Description

DETAILED DESCRIPTION OF THE INVENTION [Technical Field of the Invention] The present invention relates to a flow rate verification device for a nuclear reactor core cooling system.

[Technical Background of the Invention] In general, light water reactors are equipped with engineering technology to inject and replenish cooling water into the reactor pressure vessel when the amount of cooling water in the reactor pressure vessel decreases due to an emergency situation. An emergency core cooling system is installed as one of the safety facilities.

FIG. 1 shows a high-pressure core spray system for a boiling water reactor, which is one of such emergency core cooling systems, and in the figure, reference numeral 1 indicates a reactor pressure vessel.

A main steam pipe 2 is connected to the upper part of the reactor pressure vessel 1 to supply steam generated in the reactor pressure vessel 1 to a turbine condensate system (not shown). A normally open main steam isolation valve 3 for isolating the pressure vessel 1 is inserted. When an emergency situation occurs and the internal pressure of the reactor containment vessel (not shown) rises above a certain set pressure, or when the reactor water level drops below a set value, the condensate stored in the condensate storage tank 4 cools down. Water passes through the water supply pipe 5 and is supplied into the reactor pressure vessel 1 from the upper opening of the reactor pressure vessel 1.

A high-pressure core spray pump 6 is installed in the cooling water supply pipe 5 to transfer condensate from the condensate storage tank 4 to the reactor pressure vessel 1, and to be driven by an emergency power source even in the event of a loss of external power, for example. An on-off valve 7 (normally open), a check valve 8, and a pump suction pressure gauge 9a for measuring the suction pressure of the high-pressure core spray pump 6 are disposed upstream of the high-pressure core spray pump.

Further, downstream of the high-pressure core spray pump 6, there is a pressure gauge 9b that measures the discharge pressure of condensate from the high-pressure core spray pump 6 in order from upstream, a check valve 10, and a flow rate of condensate from the high-pressure core spray pump 6. A flow meter 11 to be measured, an on-off valve 12 (normally closed), a check valve 13, and an on-off valve 14 (usually open) are inserted.

Further, a pool water supply pipe 16 is provided which branches from the high pressure core spray pump 6 and the check valve 8 and opens into a suppression pool 15 arranged at the lower part of the reactor containment vessel (not shown).

This pool water supply pipe 16 is connected to the condensate storage tank 4
A system provided to supply pool water stored in the suppression pool 15 to the cooling water supply pipe 5 when the water level of condensate stored in the suppression pool 15 decreases or when the pool water level rises above a set value. In this pool water supply pipe 16, on-off valves 1 are installed in order from the upstream.
7 (normally closed), a check valve 18 is inserted.

In addition, an on-off valve 12 is connected to a pipe 19 and a flow meter 11 which are branched from between the pump suction pressure gauge 9a inserted in the cooling water supply pipe 5 and the check valve 10 and connected to the suppression pool 15. Pipes 20 branched from between and connected to the condensate storage tank 4 are a minimum flow line 19 and a test line 20, respectively.
This is used when testing the performance of the high-pressure core spray pump 6.

Furthermore, a test line 22 that opens into the suppression pool 15 branches off from the upstream side of the on-off valve 21 inserted in the test line 20 . And the minimum flow line 19 has an on-off valve 23 (normally closed).
However, the test line 20 has on-off valves 2 in order from the upstream.
1 (normally closed), an on-off valve 24 (normally closed) is inserted,
An on-off valve 25 (normally closed) is inserted into each of the test lines 22 . A reactor pressure gauge 26 and a suppression pool pressure gauge 28 are provided in the reactor pressure vessel 1 and the suppression pool 15, respectively.

The high-pressure core spray system configured as described above is susceptible to accidents such as, for example, an accident in which the pressure relief valve (not shown) provided in the main steam pipe 2 becomes stuck open or closed, or a pipe breaks and cooling water leaks. It is activated when the amount of cooling water in the reactor pressure vessel 1 decreases due to an accident, resulting in insufficient core cooling, which may cause the core temperature to rise and cause core damage.

In other words, in such a case, reactor pressure vessel 1
The high-pressure core spray pump 6 is activated by a water level drop signal from a water level gauge (not shown) installed in the water level gauge or a pressure high signal from the reactor containment vessel (not shown), and the on-off valve provided in the cooling water supply pipe 5 is activated. 12 is opened, condensate is supplied from the condensate storage tank 4 into the reactor pressure vessel 1, and the reactor core is cooled.

When the water level of the condensate stored in the condensate storage tank 4 falls below a certain set value or the water level of the suppression pool 15 rises, the on-off valve 17 is automatically opened and the water inside the suppression pool 15 is opened. At the same time as the pool water is supplied to the reactor pressure vessel 1, the on-off valve 7 is closed.In the high-pressure core spray system configured in this way, a predetermined amount of cooling water is supplied into the reactor pressure vessel 1. Whether or not the cooling water is being supplied is constantly monitored by an operator based on the measured value of a flow meter 11 provided in the cooling water supply pipe 5.

However, in the high pressure core spray system configured as described above, for example, the test line 20
When the on-off valves 21 and 24 or 25 respectively provided in the atomic pressure vessel 1 are open, the cooling water will not be injected into the nuclear pressure vessel 1 but will flow into the condensate storage tank 4 or the suppression pool 15.
Since the flowmeter 1 measures these flow rates, the operator determines that cooling water is flowing into the reactor pressure vessel 1 even though it is not actually flowing. That is, for example, if a closing signal is erroneously output even though the on-off valve 25 is actually open, an erroneous determination may occur. In such a case, if the on-off valve 21, 24 or 25 is not immediately closed, the accident may escalate.

Additionally, the flow meter 11 and pressure gauge 9b are out of order.
If the actual flow rate and discharge pressure are not displayed correctly, it will be extremely difficult to confirm the operating status of this high-pressure core spray system.

[Object of the Invention] The present invention has been made in response to the conventional situation, and it verifies the flow rate flowing into the reactor pressure vessel, and when there is an error in the opening/closing status signal of the opening/closing valve, etc. It is an object of the present invention to provide a flow rate verification device for a nuclear reactor core cooling system that is capable of outputting an abnormal location and also outputting a verification flow rate at this time.

[Summary of the Invention] A flow rate signal from a flow meter that measures the flow rate of cooling water flowing through a nuclear reactor core cooling system, a pressure signal from a reactor pressure system that measures the pressure in a reactor pressure vessel, and the reactor core cooling system. an input device into which a water head signal indicating the water head of a pump disposed in the system is input; and an inflow into the reactor pressure vessel based on the flow rate signal, pressure signal, and water head signal input to the input device. The expected cooling water flow rate Q ₀ ,
Calculate Q ₁ and Q ₂ and use these flow rate values Q ₀ ,
It is determined whether an abnormal situation has occurred based on the correlation between Q ₁ and Q ₂ , and if an abnormal situation has occurred, the location of the abnormality and the cooling water that is thought to have actually flowed into the reactor pressure vessel are determined. This is a flow rate verification device for a nuclear reactor core cooling system, comprising a verification device that calculates a flow rate of a nuclear reactor core cooling system, and an output device that displays an abnormal location and a flow rate of cooling water calculated by the verification device.

[Embodiment of the Invention] The details of the present invention will be described below with reference to an embodiment shown in the drawings.

FIG. 2 shows a flow rate verification device for a nuclear reactor core cooling system according to an embodiment of the present invention, which verifies the flow rate of the high-pressure core spray system described in FIG. is composed of an input device a, a verification device b, and an output device c.

The input device a inputs operating signals and the pressure inside the reactor pressure vessel 1 from various devices provided in the high-pressure core spray system shown in FIG.

That is, this input device a includes on-off valves 7 and 1.
2, 14, 17, 21, 23, 24, 25, a signal indicating the flow rate value from the flow meter 11, a signal indicating the discharge and suction pressure values from the pressure gauges 9a, 9b, high pressure core spray A signal indicating the operating state from the pump 6, a signal indicating the pressure value from the reactor pressure gauge 26 and the suppression pool pressure gauge 28, etc. are input.

The verification device b constantly monitors the state of the high-pressure core spray system to ensure that it functions normally based on various operating signals input to the input device a, and in the event of an abnormal situation, the output device described later is activated. c.
This will be indicated on the screen.

That is, this verification device b is composed of an electronic computer, processes the input signal input to the input device a using logical formulas and calculation formulas described later, and sends signals from the high pressure core spray system to the reactor reactor to the output device c, described later. The verification result of the flow rate of cooling water flowing into the pressure vessel 1 is output. Further, if it is determined that there is an error in the signal indicating the open/closed state of the on-off valve, this is output.

The verification device b will be described in detail below.

In addition to the flow rate value Q ₀ inputted from the flow meter 11, this verification device b uses flow rate values Q ₁ and Q ₂ calculated from the reactor pressure and pump water head.

In other words, there is generally a relationship between the reactor pressure input from the reactor pressure gauge 26 and the pump flow rate as shown by curve d in FIG. Flow rate Q ₁ is calculated based on the reactor pressure input from 26.

Generally, there is a relationship between the pump water head and the flow rate of the high-pressure core spray pump 6 as shown by the curve e in FIG. The pump water head is determined as the differential pressure between the pressure values, and the flow rate Q ₂ is calculated from this pump water head value.

In this way, the verification device b is used as the verification flow rate.
Three types of signals can be obtained: Q ₀ , Q ₁ and Q ₂ . However, among these three types of verification flow rates, it is calculated based on the pressure value of the reactor pressure gauge 26.
The value of Q ₀ is such that the possibility of an erroneous signal is relatively low because the reactor pressure gauges 26 are multiplexed and arranged in the reactor pressure vessel 1.

On the other hand, the pressure gauge 9b, the pump suction pressure gauge 9a, the flow meter 11, and the valve position signals are not multiplexed, and the possibility of outputting erroneous signals is relatively high. Therefore, this verification device b is the reactor pressure gauge 26.
The system is tested using the value of Q ₁ calculated from the pressure value as the most reliable parameter. That is, the verification device b has flow rates Q ₀ , Q ₁ and Q ₂
The respective flow rates are determined as follows from the correlation.

(1) Q ₀ Q ₁ Q ₂ That is, in this case, the three parameters Q ₀ , Q ₁ and Q ₂ match within the allowable error range, and the verification device b judges the system to be normal, Output Q ₀ to the output device c as the system normality and verification flow rate.

(2) Q ₀ /Q ₁ Q ₂ In this case, verification device b determines that a malfunction has occurred in flowmeter 11, and verification device b determines that flowmeter 11 has malfunctioned.
_Q2 is output to the output device c as a verification flow rate together with a failure signal.

(3) Q ₀ Q ₁ /Q ₂ In this case, verification device b determines that a malfunction has occurred in pressure gauge 9b, and verification device b
Outputs Q ₀ as a verification flow rate along with a failure signal.

(4) Q ₀ Q ₂ /Q ₁ In this case, both the flow meter 11 and the pressure gauge 9b are considered to be normal, so there is a possibility of a valve line-up error.

Therefore, the verification device b inputs the indicated value P ₃ of the suppression pool pressure gauge ₂₈ as the reactor pressure into the graph shown in FIG _. Compare with the value of Q ₂ . Then, compare this Q ₃ with the values of Q ₀ and Q ₂ and if the values almost match, it is possible that the on-off valve 25 is giving an erroneous signal of closing and is actually open. Therefore, the verification device b outputs Q ₀ as the verification flow rate along with the possibility of an erroneous signal from the on-off valve 25.

In this case, it is possible that the on-off valve 21 and the on-off valve 24 to the condensate storage tank 4 are open, but it is unlikely that both of them will issue an erroneous signal at the same time, so it is possible that the on-off valve 25 will cause an erroneous signal. The gender is output.

(5) Q ₀ /Q ₁ /QB/Q ₃ In this way, when the values of the three parameters Q ₀ , Q ₁ and Q ₂ are all different from each other, the following cases can be considered.

First, there is a case where the on-off valve 23 inserted in the minimum flow rate line 19 gives an erroneous signal indicating that it is closed, but is actually open. The mutual relationship between the values of Q ₁ , Q ₁ and Q ₂ at this time is Q>Q ₁ >Q ₀ . Here Q ₁
is larger than Q ₀ because the flow rate calculated based on the reactor pressure value input from the reactor pressure gauge 26 does not take into account leakage from the minimum flow line 19. Moreover, the reason why Q ₂ is larger than Q ₁ is because leakage occurs downstream of the pressure gauge 9b.

Further, such a relationship Q ₂ >Q ₁ >Q ₀ also occurs when the performance of the high-pressure core spray pump 6 deteriorates or the number of rotations becomes insufficient. Here, Q ₁ is larger than Q ₀ because Q ₁ has high pressure core spray pump 6.
This is because the performance deterioration of is not taken into account. Also, the reason why Q ₂ is larger than Q ₁ is due to the fourth
As is clear from the figure, the decrease in pump head pressure due to performance deterioration of the high-pressure core spray pump 6 is due to the flow rate.
This is because the value of Q ₂ will be evaluated greatly.

That is, when Q ₂ >Q ₁ >Q ₀ as described above, the two cases described above can be considered, and the verification device b outputs these two cases and Q ₀ as the verification flow rate. On the other hand, if the rotational speed of the high-pressure core spray pump 6 abnormally increases, Q ₀ > Q ₁ > Q ₂ , and in this case, the verification device b detects that the rotational speed of the high-pressure core spray pump 6 is abnormal. Output the verification flow rate Q ₀ to the output device c.

In addition, in the above comparison, the error width Q ₀ −Q ₁ , Q ₁
−Q ₂ shall be within appropriate limits.

Furthermore, a case in which the three parameters Q ₀ , Q ₁ , and Q ₂ are all different from each other in this way can be considered when the flow meter 11 fails at the same time as the pressure gauge 9b or the pump suction pressure gauge 9a.

Figure 5 shows the relationship between Q ₀ , Q ₁ , and Q ₂ at this time, divided into six cases. Cases 1 and 6 have a common pattern with the two cases described above.
Furthermore, cases 2 and 5 and cases 3 and 4 have the same failure pattern.

In this way, if Q ₀ , Q ₁ , and Q ₂ are all different from each other, six combinations are possible depending on the size, but the one caused by a single failure is the on-off valve 23 as in Cases 1 and 6. failure and performance deterioration of the high-pressure core spray pump 6, or overload of the high-pressure core spray pump 6. Therefore, in cases 1 and 6, the single fault case is considered first, and the information described above is provided, and the double fault shown in FIG. 5 is provided as a secondary one.

Furthermore, even if there is a combination of failures in the on-off valve 23 or 21 and the pressure gauge 9b or flow meter 11,
A pattern as shown in FIG. 5 can be considered. However, the possibility of such double or triple failures occurring is very low, and it is difficult to verify the flow rate, so the verification device b uses the correct flow rate.
It outputs Q ₁ to output device c, and also outputs the possibility that multiple failures have occurred and the values of various parameters to output device c, and indicates that the reactor water level will be secured depending on the system. Guide to cancel the operation and use alternative means such as monitoring reactor water level gauges.

FIG. 6 is a flowchart of the testing device b described in detail above.

In the reactor core cooling system flow rate verification device configured as described above, the high pressure core spray system is monitored as described below.

That is, the input device a includes on-off valves 7, 12, 14,
17, 21, 23, 24, 25, signals indicating the flow rate value from the flow meter 11, pressure gauge 9b, signal indicating the discharge pressure value from the pump suction pressure gauge 9a, high pressure core spray pump 6 A signal indicating the operating state from the suppression pool pressure gauge 28 and a pressure signal from the reactor pressure gauge 26 are input, and when these values are input to the verification device b, the verification device b operates as shown in FIG. The high-pressure core spray system is sequentially monitored based on the flowchart shown in , and the flow rate flowing into the reactor pressure vessel 1 is verified, and if there is an error in the open/close status signal of the open/close valve, the output is It outputs the abnormal location along with the abnormality signal to device c, and also outputs the verified flow rate value at this time. Therefore, the operator can quickly operate the abnormal area to normal operation based on the abnormal signal, abnormal area, and flow rate value displayed on the output device c, and furthermore, based on this information, Other operations can be performed quickly.

[Effects of the Invention] As described above, according to the flow rate verification device for a nuclear reactor core cooling system of the present invention, when there is an abnormality in the operating state of equipment arranged in the high-pressure core spray system, for example, an abnormality signal and The abnormal location is displayed on the display device, and the flow rate flowing into the reactor pressure vessel is displayed on this output device.
Based on these displays, the operator can quickly correct the abnormal location and perform other operations if necessary, thereby minimizing the spread of the accident.

In the embodiments described above, an example was explained in which the flow rate verification device for a nuclear reactor core cooling system of the present invention was applied to a high-pressure core spray system, but the present invention is not limited to such embodiments, and can be applied to similar systems. Of course, the present invention can be applied to a nuclear reactor core cooling system having a core cooling system.

[Brief explanation of the drawing]

Fig. 1 is a piping system diagram showing an embodiment of the high-pressure core spray system, Fig. 2 is a block diagram showing an embodiment of the flow rate verification device for the reactor core cooling system of the present invention, and Fig. 3 is a reactor pressure Figure 4 is a graph showing the relationship between flow rate and pump head; Figure 5 is an explanatory diagram showing flow rate comparison patterns; Figure 6 is the verification device shown in Figure 2. This is a flowchart. a...Input device, b...Verification device, c...Output device.

Claims (1)

[Scope of Claims] 1. A flow rate signal from a flow meter that measures the flow rate of cooling water flowing through the reactor core cooling system, a pressure signal from the reactor pressure system that measures the pressure in the reactor pressure vessel, and the reactor. an input device into which a water head signal indicating the water head of a pump disposed in the reactor core cooling system is input; Calculate the flow rate values of cooling water Q ₀ , Q ₁ and Q ₂ that are expected to be flowing into the
It is determined whether an abnormal situation is occurring based on the correlation between Q ₀ , Q ₁ , and Q ₂ , and if an abnormal situation occurs,
a verification device that calculates the location where an abnormality has occurred and the flow rate of cooling water that is thought to actually be flowing into the reactor pressure vessel; and an output device that displays the location of the abnormality and the flow rate of the cooling water calculated by the verification device. A flow rate verification device for a nuclear reactor core cooling system, comprising: and.